Diagnostic nuclear techniques generally involve the use of highly penetrating radiation comprised of nuclear particles to identify a concealed or unknown target material by detecting and measuring the interaction between the nuclear particles of the penetrating radiation and the target material nuclei, and analyzing the absorptive and/or scattering patterns that result from the interaction. For example, Thermal Neutron Activation (TNA) and Fast Neutron Activation (FNA) are neutron transmission techniques for identifying a target material by measuring the spectrum of gamma rays emitted from the target material, as a result of neutron bombardment and the subsequent absorption of neutrons by the target material. Thermal and Fast Neutron Activation methods are characterized by the energy level of the interrogating neutron beam, i.e., TNA uses a neutron beam having a low energy of about 0.025 eV, while FNA involves a very high energy neutron beam of about 14 MeV.
Although a beam of penetrating radiation transmitted through a target material may interact with the target material nuclei to produce identifiable signatures, a number of the nuclear particles comprising the penetrating radiation may also have no interaction with the target material nuclei, such that they pass through the target material, maintaining their initial speed and trajectory. The intensity of the non-interacting or uncollided radiation flux exiting the target material is less than the initial beam intensity, diminished by the fraction of interacting nuclear particles, either absorbed or scattered. For a neutron transmission, the measurement of the number of uncollided neutrons per unit time is quantified as the uncollided neutron flux or intensity, I. Analogously, the detection and measurement of the number of non-interacting photons per unit time, resulting from interrogation of the target material by a gamma-ray beam, is quantified as the uncollided gamma-ray flux or intensity, I.
With respect to neutron transmission measurements, each chemical element has a microscopic parameter, referred to as the neutron cross-section, that represents the probability of a neutron interaction with a nucleus of the target material, depending upon the velocity of the neutron. The sum of the cross-sections at a given neutron velocity is the total microscopic neutron cross-section, expressed in units of effective target area per atom (cm.sup.2 /atom). The corresponding macroscopic neutron cross-section (cm.sup.2 /cm.sup.3) is the product of the total microscopic cross-section (cm.sup.2 /atom) and the atomic density of the target material (atoms/cm.sup.3). For a chemical compound composed of n different elements, the macroscopic neutron cross-section .SIGMA..sub.C is described by Equation (1), and Equation (2) is an expression of the molecular density for a compound, C. ##EQU1##
where .SIGMA..sub.C is the total macroscopic neutron cross-section for compound C;
N.sub.C is the molecular density, molecules of compound C/cm.sup.3 ; PA1 N is Avogadro's number (6.02e23 molecules/gm mole); PA1 .nu..sub.i is the number of atoms of element i per molecule of compound C; and PA1 .sigma..sub.i is the total microscopic neutron cross-section for element i at a given neutron velocity. PA1 S.sub.o is the source strength of neutrons per second, at a certain velocity, emitted in all directions; ##EQU4## PA1 e.sup.-.SIGMA..sup..sub..SIGMA.jXj is the fraction of uncollided neutrons incident on the detector area A.sub.d.
Tables I and II below list physical characteristics of six chemical agents and their neutron kinetic energy dependence (1/2 MV.sup.2), respectively.
TABLE I Physical Characteristics of Selected Chemical Agents Mass Chemical Molecular Attenuation Agent Type Formula Weight Density Coefficient GB Nerve C.sub.4 H.sub.10 O.sub.2 PF 140.1 1.09 0.159 GA Nerve C.sub.5 H.sub.11 O.sub.2 PN.sub.2 162.3 1.07 0.156 GD Nerve C.sub.7 H.sub.16 O.sub.2 PF 182.2 1.02 0.151 VX Nerve C.sub.11 H.sub.26 O.sub.2 PSN 267.4 1.008 0.152
TABLE II Neutron Kinetic Energy Dependence of Selected Chemical Agents Energy .sigma..sub.GA .SIGMA..sub.GA .sigma..sub.GB .SIGMA..sub.GB .sigma..sub.GD .SIGMA..sub.GD .sigma..sub.VX .SIGMA..sub.VX keV cm.sup.2 /mol cm.sup.2 /cm.sup.3 cm.sup.2 /mol cm.sup.2 /cm.sup.3 cm.sup.2 /mol cm.sup.2 /cm.sup.3 cm.sup.2 /mol cm.sup.2 /cm.sup.3 150 0.596 0.63772 0.609 0.66381 0.728 0.74256 0.789 0.79689 220 0.513 0.54891 0.528 0.57552 0.63 0.6426 0.677 0.68377 272 0.475 0.50625 0.525 0.57225 0.596 0.60996 0.643 0.64943 350 0.433 0.46331 0.464 0.50576 0.541 0.55182 0.559 0.56459 450 0.477 0.51039 0.511 0.55699 0.557 0.56814 0.555 0.56055 550 0.346 0.37022 0.382 0.39458 0.428 0.43656 0.451 0.45551 660 0.324 0.34666 0.334 0.36408 0.393 0.40086 0.415 0.41915 750 0.302 0.32314 0.319 0.34771 0.373 0.38048 0.391 0.39491 850 0.283 0.30281 0.299 0.32591 0.35 0.357 0.367 0.37067 1000 0.31 0.3317 0.323 0.35207 0.358 0.36516 0.366 0.36968
For distance .delta.x within compound C, the probability of a neutron-nucleus collision is .SIGMA..delta.x, provided that this product is much less than one (&lt;&lt;1.0). The probability that a neutron travels distance X.sub.C without undergoing an interaction is a Posison distribution described by e.sup.-.SIGMA.cXc. If a neutron beam passes in series through m sucessive compounds, then the probability of zero neutron-nucleus interactions occurring over the total path length is described by Equation (3). ##EQU2##
Finally, the uncollided neutron intensity, I, is represented in the attenuation equation, Equation (4). ##EQU3##
where .epsilon. is the detector efficiency at the original neutron velocity, representing the fraction of uncollided neutrons at a certain velocity that produce counted pulses;
is the solid angle subtended by a detector of sensitive area A.sub.d located at a distance R from the point source, representing the fraction of S.sub.o neutrons leaving the source in a direction toward the detector; and
An analogous attenuation equation describes the uncollided gamma-ray intensity from gamma-ray transmissions, except that the variable .SIGMA..sub.C, the total macroscopic neutron cross-section for compound C, is replaced by the variable .mu..sub.C, the linear attenuation coefficient for compound C. The gamma-ray linear attenuation coefficient is a function of the compound's molecular structure and the gamma photon energy.
Importantly, where the parameters of Equation (4) are known, a measurement of the uncollided neutron intensity, I, allows calculation of the macroscopic neutron cross-section for the compound, .SIGMA..sub.C. Since the macroscopic neutron cross-section, .SIGMA..sub.C, is unique to each compound, the compound is identifiable by the calculation of .SIGMA..sub.C. Alternatively, measurement of the uncollided gamma-ray beam intensity, I, allows calculation of the linear attenuation coefficient, .mu..sub.C, an identifying characteristic unique to each compound.
Unfortunately, defining the parameters of the attenuation equation, Equation (4), to enable calculation of .SIGMA..sub.C or .mu..sub.C is difficult, since the parameters are largely dependent on geometry. In addition, where the target material, for example, is a chemical agent contained within a chemical munition made of a thick steel shell, the use of neutron transmission measurements to identify the chemical agent becomes more complex, and the sensitivity of the gamma radiation to the chemical agent sharply decreases, as high energy gamma-ray beams are required to penetrate the steel shell.
A need exists in the art for a safe nuclear diagnostic technique that non-intrusively identifies contained materials without relying on the geometry of the system. Furthermore, it is desirable for the nuclear diagnostic technique to be non-complex, non-time consuming, highly sensitive, reliable, and accurate, with an increased probability of correct detection and a decreased probability of false alarms.
The present invention is an improved nuclear diagnostic method for identifying a contained target material, involving the steps of measuring on-axis, mono-energetic uncollided radiation transmitted through a target material for two penetrating radiation beam energies, and applying specially developed algorithms to estimate a ratio of macroscopic neutron cross-sections for the uncollided radiation at the two energies, where the penetrating radiation is a neutron beam, or a ratio of linear attenuation coefficients for the uncollided radiation at the two energies, where the penetrating radiation is a gamma-ray beam. Alternatively, the measurements are used to derive a minimization formula based on the macroscopic neutron cross-sections for the uncollided radiation at the two neutron beam energies, or the linear attenuation coefficients for the uncollided radiation at the two gamma-ray beam energies. A candidate target material database, including known macroscopic neutron cross-sections or linear attenuation coefficients for target materials at the selected neutron or gamma-ray beam energies, is used to approximate the estimated ratio or to solve the minimization formula, such that the identity of the chemical agent is discovered.
A feature of the present nuclear diagnostic technique is that the method is performed independent of the geometry of the material containment system, such that the successive path lengths of the radiation transmissions through various compounds comprising the target material is not a factor in the identification process. The improved method accomplishes accurate and reliable measurements, providing identification within minutes.
Therefore, in view of the above, a basic object of the present invention is to provide an improved nuclear diagnostic technique for non-intrusively identifying contained materials, e.g., chemical agents, irrespective of the individual path lengths of the penetrating radiation through the container and the enclosed chemical agent.
Another object of the present invention is to provide an improved neutron transmission method for identifying a contained material by estimating a macroscopic neutron cross-section ratio for the contained material, using on-axis, uncollided neutron transmission measurements at two neutron beam energies.
Another object of the present invention is to provide an improved neutron transmission method for identifying a contained material by solving a minimization formula based on macroscopic neutron cross-sections for the contained material, using on-axis, uncollided neutron transmission measurements at two neutron beam energies.
Another object of the present invention is to provide an improved gamma-ray transmission method for identifying a contained material by estimating a linear attenuation coefficicent ratio for the contained material, using on-axis, uncollided photon measurements at two gamma-ray beam energies.
Another object of the present invention is to provide an improved neutron transmission method for identifying a contained material by solving a minimization formula based on linear attenuation coefficients for the contained material, using on-axis, uncollided photon measurements at two gamma-ray beam energies.
Yet a further object of the present invention is to provide an improved and environmentally safe nuclear diagnostic technique that has a high probability of detection and a low probability of false alarms, a simplified design, a rapid response time, an improved signal-to-noise ratio, and discrete emissions of radiation.
Additional objects, advantages, and novel features of the invention will be more fully understood from the following description of the invention, and/or will become apparent to those skilled in the art upon examination of the following description and/or by practice of the invention.